High-temperature oxidation failure in reactivity-initiated accidents; An Evaluation of failure criteria based on oxygen concentration from the previous NSRR experiments

Luu, V. N.; 谷口 良徳; 宇田川 豊; 勝山 仁哉

Nuclear Engineering and Design, 442, p.114222_1 - 114222_15, 2025/10

https://doi.org/10.1016/j.nucengdes.2025.114222

For near-term application, coated-Zr alloy claddings show potential for enhancing safety by providing better oxidation resistance and minimizing hydrogen absorption under design-basis accidents (DBA). This benefit could extend the burnup and operational cycles of fuel rods. In assessing safety, reactivity-initiated accidents (RIA) are considered as one of the DBA conditions. The current safety criteria for high-temperature oxidation failure, one of the failure modes linked to RIA, are defined by peak fuel enthalpy values that range from 205 to 270 cal/g. This wide variability presents challenges when attempting to generalize criteria for modified-Zr alloy claddings with superior oxidation resistance. Therefore, it may be more relevant to apply failure criteria based on embrittlement mechanisms, such as oxygen concentration in the -Zr phase. This study aimed to assess the failure based on both peak fuel enthalpy and cladding embrittlement by analyzing previous NSRR experiments conducted with conventional materials using the RANNS fuel performance code. The findings suggest that the failure criteria associated with cladding embrittlement can provide a rational evaluation of failure behavior compared to the existing criterion based on peak fuel enthalpy. The local failure criterion leading to the formation of through-wall cracks during quenching is consistent with Chung's proposal (NUREG/CR-1344): -Zr thickness of 0.9 wt% oxygen is less than 0.1 mm, and this corresponds to approximately 35% BJ-ECR.

加速器駆動システムの通常運転時の燃焼反応度測定精度に関する検討

方野 量太; 阿部 拓海; Cibert, H.*

JAEA-Research 2024-019, 22 Pages, 2025/05

JAEA-Research-2024-019.pdf:1.03MB

マイナーアクチノイドの核変換を目的とする加速器駆動システム(ADS)は未臨界状態で運転される。ADSの未臨界度管理においては、燃焼反応度の予測が重要であるが、予測精度の検証のためには、特に第一サイクル運転時では燃焼反応度を精度良く測定する必要がある。本検討では、燃焼反応度測定手法としてCurrent-To-Flux(CTF)法に着目し、連続エネルギーモンテカルロ計算コードSERPENT2を用いて固定源燃焼計算を実施し、炉内に配置する核分裂計数管を模したタリーを用いることで、CTF法によるADS通常運転時の燃焼反応度測定のシミュレーションを実施した。シミュレーション結果から測定手法起因の燃焼反応度測定不確かさの推定を行い、燃焼期間に依らず燃焼反応度に対して10%程度のバイアスが生じ、その検出器位置依存性が体系外側で小さいことを明らかにした。

Development of a method for the determination of spontaneous fission nuclides in irradiated fuel and applicability to Pu quantification in fuel debris by dual time neutron measurements

長谷 竹晃; 小菅 義広*; 相楽 洋*; 中岫 翔; 能見 貴佳; 奥村 啓介

Progress in Nuclear Science and Technology (Internet), 7, p.41 - 46, 2025/05

https://doi.org/10.15669/pnst.7.41

This paper provides an overview of plutonium quantification in irradiated fuel including fuel debris at the Fukushima Daiichi Nuclear Power Plants, named Dual Time Measurement (DTM) method. Spontaneous fission nuclides in irradiated fuel decrease exponentially with the passage of time according to the mainly half-life of Cm-244 (half-life of about 18.11 years). By measuring neutrons two times with long time intervals, Pu-240 effective mass (half-life of about 6,500 years) and Cm-244 mass can be quantified. Pu mass can be quantified by utilizing the correlation between ratio of Cm-244/ Pu-240 effective mass and Pu/ Pu-240 effective mass. The applicability of DTM method was evaluated numerically. The results show that long time interval was required to reduce the random errors. In the case that the interval between the first and second measurements is 32 years, Pu-240 effective mass and Pu can be quantified with uncertainties of 10-50% depending on the presence of water in storage canister and the burnup condition of irradiated fuel including the mixture of several burnup compositions in fuel debris.

Enhanced strength and ductility in an additively manufactured Al10SiMg alloy at cryogenic temperatures

Naeem, M.*; Rehman, A. U.*; Romero Resendiz, L.*; Salamci, E.*; Aydin, H.*; Ansari, P.*; Harjo, S.; Gong, W.; Wang, X.-L.*; 他3名*

Communications Materials (Internet), 6, p.65_1 - 65_13, 2025/04

https://doi.org/10.1038/s43246-025-00793-4

The need for lightweight materials with mechanical integrity at ultralow temperatures drives the development of advanced alloys for cryogenic use. Additive manufacturing via laser powder bed fusion (LPBF) offers a scalable way to create alloys with tailored properties. Here, we show that LPBF-processed Al10SiMg exhibits a high ultimate tensile strength (395 MPa) and uniform elongation (25%) at 15 K. These enhancements stem from grain refinement, increased geometrically necessary dislocations, and stress partitioning between the Al matrix and the stiffer Si phase, aiding strain accommodation. neutron diffraction reveals that the Si phase, with its higher yield strength, bears most of the load, while the Al matrix undergoes continuous strain hardening, extending deformation capacity. These results highlight Al10SiMg's promise for cryogenic applications such as hydrogen storage, aerospace, and quantum computing hardware.

Built-in physics models and proton-induced nuclear data validation using MCNP, PHITS, and FLUKA; Impact on the shielding design for proton accelerator facilities

elik, Y.*; Stankovskiy, A.*; 岩元 大樹; 岩元 洋介; Van den Eynde, G.*

Annals of Nuclear Energy, 212, p.111048_1 - 111048_12, 2025/03

https://doi.org/10.1016/j.anucene.2024.111048

The MCNP, PHITS, and FLUKA are general-purpose Monte Carlo radiation transport codes that are widely used for many real-world shielding problems at accelerator facilities around the world. For high beam energy and high beam current accelerator applications, neutron emission through the vacuum pipe along the reverse direction of incident proton beam is an important factor for a shielding design in order to correctly assess the dose rates for workers and the structural materials of the accelerator and handle with the waste activated by the backscattered neutron fluxes. In this work, neutron-production cross sections and thick target yield predictions from MC codes relying on physics models and nuclear data libraries are benchmarked against the experimental data, in order to assess their accuracy in predicting neutron emission and furthermore to assess the corresponding impact on shielding design. The results of this study demonstrate that the nuclear data libraries and physics models, which are not expected to give good results at lower energies ( MeV) but are used anyhow when there is no nuclear data available or above the energy range where the data tables end in the so-called "mix-and-match" strategy, need further improvements. Among the investigated proton induced nuclear data libraries, JENDL-4.0/HE produces the most satisfactory agreement to experimental data for all target materials, but may still benefit from refinement. Concerning the physics models of the codes, FLUKA V4-4.0 has the best performance in terms of reproducibility of the experimental values. It is also shown that all discrepancies between the calculations and the experiments for the energy range MeV are up to factor of two. This might be considered as an acceptable figure as it is equivalent to a normal safety margin () considered in shielding calculations of accelerator facilities around the world.

Measurements of neutron capture cross-section for nuclides of interest in decommissioning (II); Fe(n,)Fe

中村 詔司; 芝原 雄司*; 遠藤 駿典; Rovira Leveroni, G.; 木村 敦

Journal of Nuclear Science and Technology, 62(3), p.300 - 307, 2025/03

https://doi.org/10.1080/00223131.2024.2427244

As nuclear facilities are dismantled in decommissioning, large amount of various waste is generated. Even more inconveniently, waste is radioactive due to neutron activation. Thus, the neutron capture cross-sections of nuclides targeted in decommissioning are required to evaluate the radioactivity produced. In this work, Fe nuclide was selected among objective nuclides for decommissioning, and its thermal-neutron capture cross-section was measured by a neutron activation method at the graphite thermal column of Kyoto University Research Reactor in 5-MW operation. The thermal-neutron capture cross-section was derived using s convention. The present work obtained 1.360.03 barns for the Fe(n,)Fe reaction. The present result supports the JENDL-5 evaluation within 2. If updated with currently recommended nuclear data, some of the reported past data would support the present result.

Directional vector-based quick evaluation method for protective plate effects in X-ray fluoroscopy (DQPEX)

檜作 響子*; 藤淵 俊王*; Han, D.*; 荒川 弘之*; 古田 琢哉

Radiological Physics and Technology, 18(1), p.196 - 208, 2025/03

https://doi.org/10.1007/s12194-024-00873-z

X線透視業務に携わる医療従事者の放射線防護対策一つとして放射線防護板の使用があり、効率的な防護板による遮蔽を検討するためのシミュレーションツールの開発が求められている。モンテカルロシミュレーションはX線とX線室内に存在する様々な物質との相互作用を非常に精度良く計算できる一方で、計算時間がかかり過ぎるという欠点があるため、放射線防護板の影響を短時間で見積もる目的に、事前計算された方向ベクトルを用いて短時間に線量分布を計算する簡易線量分布計算法を新たに開発した。まず、放射線防護板が存在しない状態でのX線室内の線量分布をモンテカルロシミュレーションコードPHITSを用いて計算し、X線室内をグリッドに分割した線量分布と各位置における方向ベクトル分布を取得する。開発した簡易線量分布計算法では、グリッド位置における線量はその位置での方向ベクトルの指し示す方向からの放射線により全て付与されるものと仮定し、方向ベクトルを逆方向にトレースした方向に放射線防護板が存在する場合には、そのグリッド位置での線量値は防護板による遮蔽効果を受けるものとして線量分布を計算する。この手法を用いることで、効率的な防護板の配置を検討する上で十分な精度を保ちつつ、PHITSで全計算を実行するのに比べて、計算時間を1/6000に短縮することに成功した。

Two-group drift-flux model for upward cap-bubbly two-phase flows in large square channels

孫 昊旻; 日引 俊*

International Journal of Heat and Mass Transfer, 237, p.126445_1 - 126445_14, 2025/02

https://doi.org/10.1016/j.ijheatmasstransfer.2024.126445

Various bubbles exist in two-phase flows. A practical approach is classifying the bubbles into two groups based on their drag coefficients. Two-group two-fluid model can potentially provide the most accurate analysis of two-phase flows. Two-group drift-flux model should be established as a constitutive equation to simplify the two-group two-fluid model for its practical use. The drift-flux model for large square channels has seldom been investigated, even though such channels exist in various engineering systems. This study developed the two-group drift-flux model for large square channels based on experimental databases.

Study on the effect of radiation-resistant resin on water radiolysis

伊藤 辰也; 永石 隆二; 桑野 涼*; 神戸 正雄*; 吉田 陽一*

Radiation Physics and Chemistry, 226, p.112198_1 - 112198_5, 2025/01

https://doi.org/10.1016/j.radphyschem.2024.112198

近年、放射線照射実験や非密封放射性同位元素実験用の容器として、ポリイミドやポリエーテルエーテルケトンなどの耐放射線性樹脂の使用が増えている。しかし、水の分解生成物と樹脂との相互作用の可能性が見つかり、樹脂が放射線場における水中の反応に影響を及ぼすことが懸念された。その相互作用を明らかにするために、線による水の放射線分解における重クロム酸イオン(CrO)の還元と過酸化水素(HO)の生成を樹脂の有無で比較した。同一線量で樹脂ありの水溶液中のCrO還元量は樹脂なしの場合よりも大きくなり、樹脂によってCrO還元が促進されることを示した。一方、電子捕捉剤の有無にかかわらず、純水中のHO生成は樹脂の有無にほとんど依存しなかった。これらは、放射線場において水と接触した樹脂とヒドロキシルラジカルとの相互作用を示唆している。

Integrity evaluation of boundary function of main components in nuclear plants during severe accidents

月森 和之; 矢田 浩基

Journal of Pressure Vessel Technology, 147, p.031901_1 - 031901_9, 2025/00

https://doi.org/10.1115/1.4067760

日本では、福島第一原子力発電所の事故以来、原子力プラントに対して厳しい安全対策が取られている。シビアアクシデント時において、放射性物質を内包する危機がそのバウンダリ機能を維持できるか否かが重要な関心事となる。本研究では、高速炉の1時冷却材を内包する容器等のバウンダリを構成する構造部材である鏡板とベローズに着目し、設計を超える過大な圧力を受けた場合の座屈、座屈後変形、さらにバウンダリの貫通破損(バウンダリ機能喪失)までの挙動を検討した。研究は、2013年度に始まり、段階的に進められたが、最終段階として、新たに提案する破損クライテリアの鏡板、ベローズへの適用結果を示すものである。